From the point of view of an industrial fusion reactor, another challenge lies in the ability to ensure that the machine is self-sufficient in tritium, thereby closing the ‘fuel cycle’ internally. Because of its short half-life (12.3 years), tritium is a very rare element on Earth: natural production is 0.2 kilograms per year and the current inventory of tritium that can be mobilised on Earth is around 19 kilograms (mainly produced in CANDU-type heavy water power plants)
The tritium used in a fusion power plant will therefore have to be produced by the plant itself. The advanced technical solution is the development of a tritium blanket, positioned opposite the plasma. The deuterium-tritium (D-T) fusion reaction releases a high-energy neutron and a helium atom. While the plasma remains confined by the tokamak’s magnetic fields, the neutrons escape and are absorbed by this blanket made up of “modules” lining the wall. The presence of lithium in these blanket modules triggers another reaction: the incident neutron is absorbed by the lithium atom, generating an atom of tritium and an atom of helium. The tritium can then be extracted from the blanket and recycled in the plasma as fuel.
The challenge of tritium self-sufficiency is clear: a tritium consumed by a fusion reaction gives off a neutron that must produce new tritium. The tritium blanket will therefore have to incorporate, in addition to lithium, a neutron multiplier (such as lead or beryllium) to compensate for the inevitable loss of neutrons. It will also not be possible to cover the entire inner wall of the torus with blanket modules, if only to make room for the plasma heating devices.
Once the fusion reaction has started in a fusion reactor, all it needs to maintain it is deuterium and lithium, two elements that are abundantly available.
The possibility of producing tritium in situ in this way has now been validated at concept level. We are moving towards two types of cover: concepts using solid tritium-bearing materials, in which the tritium is extracted by circulating a gas between lithium-based ceramic balls, or concepts using liquid tritium-bearing materials, in which the liquid circulates inside the reactor during the reaction and is then treated to extract the tritium. We now need to find the best compromise, depending on the compatibility of the materials and the acceptable operating windows (operating temperature, resistance to swelling, etc.).
The choice of structural materials for the first wall and tritiated covers of fusion reactors is crucial, as it will largely determine the reactor’s performance in terms of efficiency (maximum permissible operating temperature) and availability (lifetime of the cover elements).
On this issue, the ITER programme itself will enable a choice of technological options to be put forward. To this end, tritium blanket modules of different concepts will be tested, without the mission of re-supplying the plasma with tritium, since they will occupy only a small fraction of ITER’s inner wall. The integrated demonstration of tritium self-sufficiency will be one of DEMO’s missions.
